ML21294A158
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SSES-FSAR 1.3 COMPARISON TABLES 1.3.1 COMPARISONS WITH SIMILAR FACILITY DESIGNS This subsection highlights the historical principal design features of the plant and compares its major features with other boiling water reactor facilities. The design of this facility was based on proven technology obtained during the development, design, construction, and operation of boiling water reactors of similar types.
The data, performance, characteristics, and other information presented here represent the then current Susquehanna Steam Electric Station design as it compared to similar designs available at that time. To maintain this original design comparison, these tables will not be revised to reflect current plant design, other than the previous addition of the fifth ("E") emergency diesel generator to Tables 1.3-6 and 1.3-7.
1.3.1.1 Nuclear Steam Supply System Design Characteristics Table 1.3-1 summarizes the design and operating characteristics for the nuclear steam supply systems. Parameters are related to rated power output for a single plant unless otherwise noted.
1.3.1.2 Power Conversion System Design Characteristics Table 1.3-2 compares the power conversion system design characteristics.
1.3.1.3 Engineered Safety Features Design Characteristics Table 1.3-3 compares the engineered safety features design characteristics.
1.3.1.4 Containment Design Characteristics Table 1.3-4 compares the containment design characteristics.
1.3.1.5 Radioactive Waste Management Systems Design Characteristics Table 1.3-5 compares the radioactive waste management design characteristics.
Rev. 49, 04/96 1.3-1
SSES-FSAR 1.3.1.6 Structural Design Characteristics Table 1.3-6 compares the structural design characteristics.
1.3.1.7 Instrumentation and Electrical Systems Design Characteristics Table 1.3-7 compares the instrumentation and electrical systems design characteristics.
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION All of the significant changes that have been made in the facility design between submission of the last PSAR revision and Revision 0 of the FSAR are listed in Table 1.3-8. Each item in Table 1.3-8 is cross-referenced to the appropriate portion of the FSAR which describes the changes and the bases for them.
Rev. 49, 04/96 1.3-2
SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
- HATCH 1 BWR 4
- ZIMMER BWR 5
- GESSAR BWR 6 251-764 218-560 28-560 238-732 THERMAL AND HYDRAULIC DESIGN Rated power, MWt 3293 2436 2436 3579 Design power, MWt (ECCS design basis) 3439 2550 2550 3758 Steam flow rate, lb. hr. 13.48 x 106 10.03 x 106 10.477 x 106 15.396 x 106 Core coolant flow rate, lb/hr. 100.0 x 106 78.5 x 106 78.5 x 106 105.0 x 106 Feedwater flow rate, lb/hr. 13.574 x 106 10.445 x 106 10.477 x 106 15.358 System pressure, nominal in steam dome, psia 1020 1020 1020 1040 Average power density, KW/liter 48.7 51.2 50.51 56.0 Maximum thermal output, KW/ft. 13.4 13.4 13.4 13.4 Average thermal output, KW/ft. 5.34 7.11 5.45 6.04 Maximum heat flux, Btu/hr-ft2 361,000 428,300 354,000 354,300 Average heat flux, Btu/hr-ft2 144,100 164,700 143,900 159,600 Maximum UO2 temperature, F 3330 4380 3325 3337 Average volumetric fuel temperature, F 1100 1100 1100 1100 Average cladding surface temperature, F 558 558 558 558 Minimum critical power ratio (MCPR) 1.23 1.9* 1.21 1.24
- For Hatch minimum critical heat flux (MCHFR) ws used.
Coolant enthalpy at core inlet, Btu/lb 521.8 526.2 527.4 527.9 Core maximum exit voids within assemblies 76 79 75 76 Core average exit quality, % steam 13.2 12.9 13.6 14.9 Feedwater temperature, F 383 387.4 420 420 THERMAL AND HYDRAULIC DESIGN Design Power Peaking Factor:
Maximum relative assembly power 1.40 1.40 1.40 1.40 Local peaking factor 1.15 1.24 1.24 1.13 Axial peaking factor 1.40 1.50 1.40 1.40 Total peaking factor 2.51 2.60 2.43 2.22 NUCLEAR DESIGN (First Core)
Water/U02 volume ratio (cold) 2.80 2.53 2.41 2.70 Reactivity with strongest control rod Keff F0.99 F0.99 F0.99 F0.99 Moderate void coefficient:
Hot, no voids k/k - % void -1.0 x 10-3 -1.0 x 10-3 -1.0 x 10-3 -0.3 x 10-3 At rated output, k/k - % void -1.7 x 10-3 -1.6 x 10-3 1.6 x 10-3 -1.0 x 10-5 Fuel temperature doppler coefficient:
At 68 F, k/k - F fuel -1.2 x 10-5 1.3 x 10-5 -1.3 x 10-5 -1.6 x 10-5 Rev. 55, xx/xx Page 1 of 4
SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
SSES HATCH 1 ZIMMER GESSAR BWR 4 BWR 4 BWR 5 BWR 6 251-764 218-560 28-560 238-732 Hot, no voids, k/k F fuel -1.2 x 10-5 -1.2 x 10-5 -1.2 x 10-5 -1.3 x 10-5 At rated output, k/k,F fuel -1.2 x 10-5 -1.3 x 10-5 -1.3 x 10-5 -1.2 x 10-5 Initial average U-235 enrichment wt. % 1.88 2.23 1.90 1.90 Fuel average discharge exposure, MWd/short ton 16,200 19,000 15,053 13,000 CORE MECHANICAL DESIGN (First Core)
Fuel Assembly:
Number of fuel assemblies 764 560 560 732 Fuel rod array 8x8 7x7 8x8 8x8 Overall dimensions, in. 176 176 176 176 Weight of U02 per assembly lb. (pellet type) 458 (chamfered) 490.4 (undished) 465.15 472 483.4 (dished) (chamfered)
Weight of fuel assembly, lb. 600 681 (undished) 698 675 (dished)
Fuel Rods:
Number per fuel assembly 62 49 63 63 Outside diameter, in 0.483 0.563 0.493 0.493 Cladding thickness, in 0.032 0.032 0.034 0.034 Gap, pellet to cladding, in 0.0045 0.006 0.0045 0.009 Length of gas plenum, in 10 16 14 12 Cladding material* Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2
- Free-standing loaded tubes Fuel Pellets:
Material U02 U02 U02 U02 Density, % of theoretical 95 95 95 94 Diameter, in 0.410 0.487 0.416 0.416 Length, in 0.410 0.5 0.420 0.420 Fuel Channel:
Overall dimension, length, in 166.9 166.9 166.9 Thickness, in 0.080 0.080 0.100 0.120 Cross section dimensions, in 5.48 x 5.48 5.44 x 5.44 5.48 x 5.48 5.52 x 5.52 Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Core Assembly:
Fuel Weight as UO2 lb 349,000 272,850 250,538 345,500 Core diameter, (equivalent), in 187.1 160.2 160.2 Core height (active fuel) in 150 144 146 148 Rev. 55, xx/xx Page 2 of 4
SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
SSES HATCH 1 ZIMMER GESSAR BWR 4 BWR 4 BWR 5 BWR 6 251-764 218-560 28-560 238-732 Reactor Control System:
Method of variation of reactor power Movable control rods and Movable control rods and Movable control rods and variable Movable control rods and variable variable forced coolant flow variable forced coolant flow forced coolant flow forced coolant flow Number of movable control rods 185 137 137 177 Shape of movable control rods Cruciform Cruciform Cruciform Cruciform Pitch of movable control rods 12.0 12.0 12.0 12.0 Control material in movable rods B4C granules B4C granules B4C granules B4C granules compacted in SS compacted in SS compacted in SS compacted in SS tubes tubes tubes tubes Type of control rod drives Bottom entry locking piston Bottom entry locking piston Bottom entry locking piston Bottom entry locking piston Type of temporary reactivity control for initial core Burnable poison; gadolinia- Burnable poison; gadolinia-urania Burnable poison; gadolinia-urania fuel Burnable poison; gadolinia-urania fuel urania fuel rods fuel rods rods rods Incore Neutron Instrumentation:
Number of incore neutron detectors (fixed) 172 124 124 164 Number of incore detector assemblies 43 31 31 41 Number of detectors per assembly 4 4 4 4 Number of flux mapping neutron detectors 5 4 4 Range (and number) of detectors:
Source range monitor Source to 0.001% power (4) Source to 0.001% power (4) Source to 0.001% power (4) Source to 0.001% power Intermediate range monitor 0.001% to 10% power (8) 0.001% to 10% power (8) 0.001% to 10% power (8) 0.001% to 10% power Local power range monitor 5% to 125% power (172) 5% to 125% power (124) 5% to 125% power (124) 5% to 125% power Average power range monitor 5% to 125% power (6)* 2.5% to 125% power (6)* 2.5% to 125% power (6)* 2.5% to 125% power
- Channels of monitors from LPRM detectors.
Number and types of incore neutron sources 7 Sb-Be 5 Sb-Be 5 Sb-Be REACTOR VESSEL DESIGN Material Carbon steel/Stainless Clad Carbon steel/Stainless Clad Carbon steel/Stainless Clad Carbon steel/Stainless Clad Design pressure, psig 1250 1265 1250 1250 Design temperature, F 575 575 575 575 Inside diameter, ft-in. 20-11 18-2 18-2 19-10 Inside height, ft-in. 72-11 69-4 69-4 70-10 Minimum base metal thickness(cylindrical section) in 6.19 5.53 5.375 5.70 Minimum cladding thicknesses, in 1/8 1/8 1/8 1/8 REACTOR COOLANT RECIRCULATION DESIGN Number of recirculation loops
- 2
- 2
- 2
- 2 Design pressure: - - - - -~
Rev. 55, xx/xx Page 3 of 4
SSES-FSAR Table 1.3-1 COMPARISON OF NUCLEAR STEAM SUPPLY SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
SSES HATCH 1 ZIMMER GESSAR BWR 4 BWR 4 BWR 5 BWR 6 251-764 218-560 28-560 238-732 Inlet leg, psig 1250 1148 1250 1250 Outlet leg, psig 1500 1274 1675*;1575** 1675*;1575**
- Pump and discharge piping to and including discharge block valves.
- Discharge piping from discharge block valve to vessel Design temperature, F 575 562 575 575 Pipe diameter, in 28 28 20 22/24 Pipe material, ANSI 304/316 304/316 304/316 304 Recirculation pump flow rate, gpm 45,200 42,200 33,880 35,400 Number of jet pumps in reactor 20 20 20 20 MAIN STEAMLINES Number of streamlines 4 4 4 4 Design pressure, psig 1250 1146 1250 1250 Design temperature, F 575 563 575 575 Pipe diameter, in 26 24 24 26 Pipe material Carbon Steel Carbon Steel Carbon Steel Carbon Steel Rev. 55, xx/xx Page 4 of 4
SSES-FSAR Table 1.3-2 COMPARISON OF POWER CONVERSION SYSTEM DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
- SSES
- HATCH
- ZIMMER
- GESSAR BWR 4 BWR 4 BWR 5 BWR 6 251-764 218-560 218-560 238-732 TURBINE GENERATOR (See Sections 10.2 and 10.4)
Rated power, MWt 3293 2550 2550
- Rated power, MWe (gross) 1085 813 883
- Generator Speed, RPM 1800 1800 1800
- Rated steam flow, lb/hr 13.4 x 106 10.48 x 106 11.0 x 106
- Inlet pressure, psig 965 950 950
- STEAM BYPASS SYSTEM (See Section 10.4.4)
Capacity, % design steam flow
- 25
- 25
- 25 MAIN CONDENSER (See Section 10.4.1)
Heat removal capacity, Btu/hr 7890 x 106 5720 x 106 - 7053 x 106
- CIRCULATING WATER SYSTEM (See section 10.4.5)
Number of pumps 4 2
- 3
- Flow rate, gpm/pump
- -- 112,000
- -- 185,000 - 150,000 CONDENSATE AND FEEDWATER SYSTEM Design flow rate, lb/hr 13.44 x 106 10.096 x 106 10.971 x 106
- Number of condensate pumps 4 3 3
- Number of condensate booster pumps None 3 3
- Number of feedwater pumps 3 2 2
- Number of feedwater booster pumps None None None
- Feedwater pump drive Turbine Turbine Turbine
- Feedwater booster pump drive NA NA NA *
- See applicants SAR Rev. 49, 04/96 Page 1 of 1
Table 1.3-3 COMPARISON OF ENGINEERED SAFETY FEATURES DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
Table 1.3-4 COMPARISON OF CONTAINMENT DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
Table 1.3-5 RADIOACTIVE WASTE MANAGEMENT SYSTEMS DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
SSES-FSAR Table 1.3-6 COMPARISON OF STRUCTURAL DESIGN CHARACTERISTICS (Parameters are for rated power output for a single plant unless otherwise noted. These parameters were current at issuance of the SSES Operating License and are being maintained for historical reference.)
SSES HATCH 1 ZIMMER GESSAR BWR-4 BWR 4 BWR 5 BWR 6 251-764 218-560 218-560 238-732 SEISMIC DESIGN* (See Section 3.7)
- horizontal g 0.05 0.08 0.10 0.15
- vertical g 0.033 0.05 0.07 ----
- horizontal g 0.10 0.15 0.20 0.30
- vertical g 0.067 0.10 0.14 ----
WIND DESIGN (See Section 3.3)
Maximum sustained - mph f--- 80 f--- 105 f--- 90 ~ 130 TORNADOS (See Section 3.3)
Translational - mph 60 60 60 70 Tangential - mph 300 300 300 290
- Some of the tabluated values differ for the design of the Diesel Generator E Facility.
Rev. 49, 04/96 Page 1 of 1
Table 1.3-7 COMPARISON OF ELECTRICAL POWER SYSTEM DESIGN CHARACTERISTICS Security-Related Information Table Withheld Under 10 CFR 2.390
SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
FSAR PORTION IN WHICH CHANGE IS ITEM CHANGE REASON FOR CHANGE DISCUSSED Recirculation flow measurement The recirculation flow measurement To improve flow measurement 7.3.1, 7.6.1 design was changed from a flow accuracy.
element to an elbow-tap type.
Recirculation system The pressure interlock for RHR NRC Requirement for diversity. 7.3.1, 7.6.1 shutdown mode was changed.
Nuclear fuel The number of fuel pins in each fuel Improved fuel performance by 4.2 bundle has been changed from 7 x 7 increasing safety margins.
to 8 x 8.
Nuclear boiler An additional test mode was added Verifies that the spring force on the 5.4 for closing MSIVs one at a time to valves will cause them to close 90% of full open in the fast mode under loss-of-air conditions.
(close in slow mode already existed).
Main steam line isolation A main condenser low vacuum NRC requirement 7.3.1 initiation of the main steam line isolation was added.
Main steam line isolation Reactor isolation was deleted for To provide improved plant 5.4 high water level initiation actuation. availability.
Main steam line drain system A main steam line drain system was Prevent accumulation of condensate 5.4 improved. in an idle line outboard of MSLIV.
Feedwater sparger The thermal sleeve was changed to To eliminate vibration, failure, and 5.3 provide improved design of sparger leakage.
to nozzle.
Standby liquid control (SLC) system Interlocks on the SLC system were To prevent inadvertent boron 9.3.5 and 7.4.1 revised. injection during system testing.
Rev. 49, 04/96 Page 1 of 3
SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
FSAR PORTION IN WHICH CHANGE IS ITEM CHANGE REASON FOR CHANGE DISCUSSED RCIC & HPCI steam supply A warmup bypass line and valve Permits pressurizing and pre- 5.4 and 6.3 was added. warming of the steam supply line downstream to the turbine during reactor vessel heatup.
RCIC & HPCI vacuum breaker A vacuum breaker system was To prevent backup of water in the 5.4 and 6.3 system added to the turbine exhaust line pipe and consequential high into the suppression pool. dynamic pipe loads and reactions.
RCIC & HPCI system Each component has been made Improved testability 5.4 and 6.3 capable of functional testing.
Automatic depressurization system The interlocks on the automatic To meet IEEE-279 requirements. 7.3.1 (ADS) depressurization system were revised.
RPV code The RPV was partially updated to Update to applicable code as much 5.2 ASME 1971 code and Summer 1971 as practical.
addenda.
Level instrumentation The RPV level instrumentation was Improve ECCS separation per IEEE 7.3.1 revised to eliminate Yarway columns 279 and improve reliability.
and replace them with a conventional condensing chamber type; also, separation and redundancy features were added.
Leak detection system The leak detection system was To meet IEEE-279 requirements. 7.6.1 revised to upgrade the capability.
Reactor vibration monitoring A confirmatory vibration monitoring NRC requirement 14.2 test was added.
Rev. 49, 04/96 Page 2 of 3
SSES-FSAR TABLE 1.3-8 SIGNIFICANT DESIGN CHANGES FROM PSAR TO FSAR*
FSAR PORTION IN WHICH CHANGE IS ITEM CHANGE REASON FOR CHANGE DISCUSSED Primary Containment Concrete Delineation of compressive Update to reflect current engineering 3.8B strengths for pozzolan vs. non- design requirements.
pozzlan Type II Portland cements.
RPV Insulation Correct the RPV Insulation Revised support beams on as-build 5.3.3.1.4 Description RPV Insulation Panels Safety Related Conduits & Trays Correct separation statements for Question 7.4 of Amend. #5 of PSAR 3.12 conduits and trays. (Revised per requirement of Reg.
Guide 1.75 - 1974).
Tornado Loading Revised Tornado Loading To reflect latest NRC 3.3 combinations. recommendations in the Standard Review Plan.
- NOTE: Design changes listed are only those which have occurred between the last SSES PSAR Amendment and Revision 0 of the FSAR.
The NRC has been notified of all other design changes prior to the last PSAR amendment by previous amendments to the PSAR.
Rev. 49, 04/96 Page 3 of 3